ML20003G000
| ML20003G000 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 04/16/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Parker W DUKE POWER CO. |
| References | |
| NUDOCS 8104280022 | |
| Download: ML20003G000 (8) | |
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e UNITED STATES
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NUCLEAR REGULATORY COMMISSION WASHINCTON, D. C. 20555
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APR 16 1981 Docket Nos.: 50-413/414 Mr. William O. Parker Vice President - Steam Production Duke Power Company P. O. Box 33189 Charlotte, North Carolina 28242
Dear Mr. Parker:
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SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - CATAWBA NUCLEAR STATION UNITS 1 AND 2 The Instrumentation and Control Systems Branch has identified four concerns that will be addressed in its review of operating license applications.
The specific concerns are delineated in the enclosure. We request that you amend your Final Safety Analysis Report to reflect your responses within 45 days of the date of this letter. Should you have any questions, contact our Licensing Project Manager, Kahtan Jabbour.
Sincerely.
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Robert L. Tedesco, Assistant Director I
for Licensing Division of Licensing
Enclosure:
i As stated cc: See next page.
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Mr. William 0. Parker cc-William L. Porter, Esq.
Duke Power Company P. O. Box 33189 Charlotte, North Carolina 28242 J. Michael McGarry, III, Esq.
Debevoise & Liberman 1200 Seventeenth Street, N. W.
Washington, D. C.
20036 North Carolina PPA-!
P. O. Box 95162 Raleigh, North Carolina 27625 Mr. R. S. Howard Power Systems Division Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230 Mr. C. W. Woods NUS Corporation
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2536 Countryside Boulevard Clearwater, Florida 33515 Mr. Jesse L. Riley, President Carolina Environmental Study Group 854 Henley Place Charlotte, North Carolina 28208 Richard P. Wilson, Esq.
Assistant Attorney General S. C. Attorney General's Office P. O. Box 11549 Columbia, South Carolina 29211 Walton J. McLeod, Jr., Esq.
General Counsel South Carolina State Board of Health J. Marion Sims Building 2600 Bull Street Columbia, South Carolina 29201 e.
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0 2-Mr. William 0. Parker cc: James W. Burch, Director -
Nuclear Advisory Counsel 2600 Bull Street Columbia, South Carolina 29201 Mr. George Maxwell, Resident Inspector c/o U. S. Nuclear Regulatory Commission P. O. Box 11695 Rock Hill, South Carolina 29730 e
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- 2.0 Instru.er,tation 3 Control Systa s Srcnch M 2.01 Loss of On-Class IE Instru. antatien and Cortrol Pc-er System EF OlIr??.g7 oC2TCErTiicn t IE7 illEirNu )
If reacter controls and vital instrumants derive pcaer frca ccnon electrical distribution systeas, the failure of such electrical distribution systeas may result in an event requiring operator acticn concurrent with failure of irport.nt instrurentatien upcn which these cperator actions should be b* sed. This cencerns was addressed in IE Bulletin 79-27. On N verber 30, 1979, IE Euilatin 79-27 was sent to cperating license (OL) holders, the r. ear tera CL applicants (North Anna 2, Diablo Canyon, McGuire, Salen 2, Sequoyah, and Zir er),
cad other holders of ccastructica parnits (CP), including Catawba, Of these recipients, the CP holdars were not given explicit directicn for making a submittal as part of the licensing review.
- However, they were informed that the issue would be addressed later.
Ycu are requested to address these issue by taking IE Bulletin 79-27 Actions 1 thru 3 under " Actions to be Taken by Licensees".
Within the response time called for in the attached transr.ittal letter, cc-r.plete the review and evaluaticn required by Actions 1 thru 3 and provide a written response describing ycur reviews and
- ctions. Tiiis raport should be in the for a of an a.endnent to your FSAR and su' aitted to the."RC Of fice cf tcisar Lictor v
Regulation as a licensing subnittal.
2?2.02 Engineered Safety Features (ESF) Reset Centrols (IE Pulletin 80-06)
If safety equipment does not remin in its every.ccy ide upon reset of an engineered safeguards actuation signal, syste.1 <:odification, design change or othe corrective action should be planned to assure that p otective action of the affected equipment is not comprcaised cace the associated actuation signal is raset. This issue uns addressed in IE Bulletin 80-05 (enclosed). For f.:-cili ties with cperating licenses as of March 13, 1980, IE bulletin 30-06 rcquired that reviews be conducted by the licensees to daterair.e which, if any, safety functions aight be unavailabe af'.ar rosat, and uhat changes could be implemented to correct the prcblem.
For facilities with a constructicn peruit including OL :pplic:atsBulletin 80-05 was issued for informatica caly.
The NRC staff has deter lined that all CP holders, as a part of the OL revicw process are to be requestad to address this issue.
Accordingly, you are requested to take the acticas called for in Bulletin G0-05 Actions 1 thru 4 under " Actions to be Tcken by Licensees". Within the response time called for in the attac:. d transm.ittal letter, cccplete the review verifications and descri A. ns
?00R BRIGINAL
of cc. rective actions taken or pican :d cs stated in.',ctin1 i t5ra 3 and r.tmit the rep 0rt called for in A:tions Itea 4 The rep rt should be submitted to the NRC Office of Nuclear Re;ulatica as a licensing submittal in the form of an FSAR amendment.
122. 03 Oualification of Control Systems (IE Infornation Notice 79-22) 0?erating reactor licensees were inforced by IE Infornation Notice 7?-22, issued September 19, 1979, that certain non-safety grade or control equiprent, if subjected to the adverse environ. Tent of a high energy line break, could inpact the safety analyses and the adequacy of the protecticn functions perforced by the safety grade equiprent.
Enclosed is a copy of IE Inferration Notice 79-22, a-d reprinted copies of an August 20, 1579 Westinghouse letter and a September 10, 1979 Public Service Electric and Gas Cocpany letter which address this matter. Operating Reactor licensees conducted reviews to determine whether such problems could exist at operating facilities.
W2 are concerned that a similar potential may exist at light water facilities now under construction. You are, therefore, requested to perform a review to determine what, if any, design changes or cperator actions wculd be necassary to assure that high energy line breaks will not cause centrol system failures to ccaplicate the event beyond ycur FSAR analysis.
Provide the roaults of ycur reviews including all identified problems and the manner in..hich ycu.have resolved them to NRR.
The specific " scenarios" discussed in the above referenced Westinghcuse letter are to be censidered as examples of the_ kind cf interactions which might occur. Your review should include these scenarios, where appiciable, 'aut should not nacessarily be limited to tham.
Applicants with other LWR designs should consider anales:us int:ractions as relevant to their designs.
i.22. 04 Centrol System Failures The analyses reported in Chcpter 15 of the FSAR are intended to d2=onstrate the adequacy of safety systems in mitigating anticipatad cperational occurrences and accidents.
Sas ?d cn the ccnservative asre:cpticas acde in defining these dcsi.: br.ais events and the detailed ravicw of ue u..aiyses by the staff, it i; li'. ly that they adequately bound the consequcncc3 of single control syst;m failures.
To provide assurance that the design basis rzent analyses adequately bound other more fundamental credible failures you are requested to provide the following informa tion:
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P00R ORIGINAL.
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Identify those centrol syst-,s whose failure or nalfuncticn cculd sericusly impact plant safety.
(2)
Indicate which, if any, of the centrci systems identified in (1) receive power fran cc r:n power sources. The porer sources considered shculd incl'.He all power sources whose failure or malfunction could lead to failure or calfuction cf rare than one control system and should extend to the effects of cascading power losses due to the failure of higher level distribution panels and load centers.
(3)
Indicate which, if any, of the control systems identified in (1) receive input signals froa cccmon sensors.
The. sensors considered should include, but shculd not necessarily be limited to, cer cn hydraulic headers or inpulse lines feeding pressure, temperature, level or othar signals to t-o or more control systems.
(4)
Provide justification that any simultaneous malfunctions of the control systems indentified in (2) and (3) resulting frca failures or malfunctions of the applicable common power source or senscr are bounded by the analyses in Chapter 15 and would not require action or response beycnd the capability of operators or safety systems.
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UNITED STATES SSINS No.: 6820 h0 CLEAR REGULATORY CCMMISSION Arension No.:
0FFICEOFINSPECTIONANDENFORCEMENTQ910250499 WASHINGTON, D.C.
20555 November 30, 1979 IE Bulletin No. 79-27
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_i, LOSS GF NON-CLA55-1-E INSTRUMENTATION AND CONTROL POWER SYSTEM BUS
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Cescription of Circumstances:
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Cn Novecber 10, 1979, an event occurred at the Oconee Power Station, Unit 3, L-r tnat resulted in loss of power to a non-class-1-E 120 Vac single phase power J
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panal that supplied power to the Integrated Control System (ICS) and the Non-Nuclear Instrumentation (NNI) System. This loss of power resulted w4 in control system malfunctions and significant loss of inforpation to the
.,j control rocs operator.
i Joecifically, at 3:15 p.m., with Unit 3 at 100 percent power, the main condensate
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pscas tripped, apsarently as a result of a technician performing maintenance on 19e hotwell level control system. This led to reduced feedwater flow to the stes: generators, which resulted in a reactor trip due to high coolant system At 3:17:15 p.m., the p essure and simultaneous turbine trip at 3:16:57 p.m.
r.on-class-1-E inverter power supply f eeding all power to the integrated control system (which provides proper coordination of the reactor, steam generator feat.ater control, and turbine) and to one NNI channel tripped and failed to automatically transfer its loads from the DC power source to the regulated AC The inverter tripped due to blown fuses. Loss of power to the i
j p:wer source.
!WI rendered control room indicators and recorders for the reactor coolant system
'A (ex:ept for one wide-range RCS pressure recorder) and most of the secondary plant L
sjste:s incperaale, causing loss of indication for systems used for decay heat removal anc water addition to the reactor vessel and steam generators. Upon loss
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' of power, all valves controlled by the ICS assumed their respective failure The loss of power existed for approximately three minutes, until an.
?>d positions.
coerator could reach the equipment roca and manually switch the inverter to the 1
regulated AC source.
The above event was discussed in IE Information Notice No. 79-29, issued
- Novecher 15, 1979.
SUREG 0600 " Investigation into the March 28, 1979 TMI Accident" also discusses 78-021-03L whereby the RCS depressurized and Safety Injection occured TMI LER i
on loss of a vital bus due to inverter failure.
'a Actions to Be Taken by Licensee _s M
For all power reactor facilities with an operating license and for those nearing comsletion of construction (North Anna 2, Diablo Canyon, McGuire, Salem 2, Sequoyah, and li:naer):
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Review the class-1-E and non-class 1-E buses supplying power to safety and 1.
i non-safety related instrumentation and control systems which could affect if the ability to achieve a cold shutdown condition using existing procedures,
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For each bus:
or procedures developed under item 2 below.
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i, identify and review the alarm and/or indication provided in the control 7'
a) room to alert the operator to the loss of power to the bus.
I' identify the instrument and control system loads connected to-the bus b) and evaluate the effects of loss of power to these loads including the ability to achieve a cold shutdown condition.
fl describe any proposed design modifications resulting from these ieviews 3]l c) and evaluations, and your proposed schedule for implementing those
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Prepare emergency procedures or review existing ones that will be used by 2.
control room operators, including procedures required to achieve a cold t
t shut 4cwn condition, upon loss of power to each class 1-E and non-class 1-E bus supplying power to safety and non-safety related instrument and 1 1 control systems. The emergency procedures should include:
I, the diagnostics / alarms / indicators /synptom resulting from the review
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a) and evaluation conducted per ites I above.
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the use of arternate indication and/or control circuits which may be b) powered from other non-class 1-E or class 1-E,. instrumentation and control buses. c) methods for restoring power to the bus. j Describe any proposed design modification or administrative controls to be icplemented resulting from these procedures, and your proposed schadule for implementing the changes. Re-review IE Circular No. 79-02, Failure of 120 \\'olt Vital AC Power Supp112s, y1 to include both class 1-E and non-class 1-E safety 3. dated January 11, 1979, Based on a review of operating experience related power supply inverters. and your re-review of IE Circular No. 79-02, describe any proposed design modifications or administrative controls to be implemented as a result of l trie re ' review. Within 90 days of the date of this Bulletin, complete the review and j j evaluation required by this Bulletin and provide a written response 4. i describing your reviews and actions taken in response to each item. } ~ =iJ Office and a copy should be forwarded to the NRC Office of 20555. Enforcement, Division of Reactor Operations Inspection, Washington, D.C. If you desire additional information regarding this matter, please contact the IE Regional Office. e=
. i; ( tiovesser 30, 1979 IE Bulletin No. 79-27 Page 3 of 3 Approved by GAO B180225 (R0072); cleara..ce expires 7/31/80. Approval was given under a blanket clearance specifically for identified generic problems. e 9 k e. ? e 1 N - f i 4 e I i! .}< o O 1 4 e 9
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.,-.i < : 7 ( IE Bulletin No. 79-27 Enclosure N:vebmer 30, 1979 RECENTLY ISSUED IE BULLETINS .s Bulletin Subject Date Issued - Issued To N3. y ;., 79-26 Boron loss From SWR 11/20/79 All SWR power reactor Control Blades facilities with an i OL j 79-25 Failures of Westinghouse 11/2/79 All power reactor _. 'J BFD Relays In Safety-Related facilities with an b*#" ';.a Systems OL or CP .g e .mr nl 79-17 Pipe Cracks In Stagnant 10/29/79 All PVR's with an (Rev. 1) Borated Wa er System At OL and for information d PWR Plants to uther power reactors M-24 Frozen Lines 9/27/79 All power reactor facilities which have either OLs or cps and + are in the late stage of construction 79-23 Potential Failure of 9/12/79 All Power Reactor Emergency Diesel Facilities with an Generator Field Operating License or l Exciter Transformer a construction permit .} 79-14 Seismic Analyses For 9/7/79 All Powar Reactor i (Supplement 2) As-Built Safety-Related Facilities with an l Piping Systems OL or a CP i 79-22 Possible Leakage of Tubes 9/5/79 To Each Licensee ~N of Tritium Gas in Time-Vbo Receives Tubes pieces for Luminosity of Tritium Gas ,~ Used in Timepieces for Luminosity 79-13 Cracking in Feedwater 8/30/79 All Designated (Rev. 1) System Piping Applicants for OLs 79-02 Pipe-Support Base Plate 8/20/79 All power Reactor } h-qgg (Rev. 1) Designs Using Concrete Facilities with an t y,r ; (Supplement 1) Expansion Anchor Bolts OL or a CP l 'l 79-14' Seismic Analyses For 8/15/79 All Power Reactor (Supplement) As-Built Safety-Related Facilities with Piping Systems an OL or a CP f a e ga" ---9 9-*-- w9g -%v-y r- ,--vn-w.r-- syvw-y g..y,v. a,g-q- ,e.c a e g iy.g .-------yw.-6wp-m-4-----e-- ---yr 9-.g.-+------t--T -T-rw--- r-9e7-*w--T-'t9--a*w-e
i l nitu:. Ju ~' L 1 cr :- % t'o. : (L.22:5',559 c:.iit0 diATES IUCL.C.JJ AinCRY CO:'.".:CSION OFFICE GT ~; ~..V.CI!ON MiD ENFORCE!ENT l tWI.'GTON, D. C. 20555 March 13, 1980 IE Bulletin No. 80-05 i ENGINEERED SAFETY FEATURE (ESF) RESET CONTROLS y Description of Circumstances: On November 7,1979, Virginia Electric and Power Company (VEPCO) reported that i following initiation of Safety Injection (SI) at North Anna Power Station j Unit 1, the use of the SI Reset pashhuttons alone resulted in certain ventila-i tion d:cpers changing position from their safety or emergency mode to their normal node. Further investigatica by VEPCO and the architect-engineer resulted ..,_ 1 in discovery of circuitry which similarly affected components actuated by a y, i Contain ent Depressurization Actuation (CDA, activated on Hi-Hi Containment Pressure). The ci'rcuits in question are listed below: f y ~ Component / System Probien l l Outside/Inside Recirculatio.n Spray Pump motors will not start after actuation if CDA Reset is depressed l Pump Motors prior to starting timer running out (approx. 3 minutes) Pressurized Control Room Dampers will open on SI Reset Ventilation Isolation Dampers Safeguards Arec Filter Dampers Dampers reposition to bypass l filters when CDA Reset is depressed ) ) Containment Recirculation Cooler Fans will rest' art when CDA Reset l Fans is depressed Service Water Supply and Discharge If service water is being used as Valves to Containment the cooling medium prior to CDA actuation, valves will reopen upon depressing CDA reset Service Water Radiation Monitoring Pumps will not start after i actuation if CITA reset is depressed Sacple Pumps prior to motor starting timers running out 9 :"9 qw Main Condcaser Air Ejector Exhaust After receiving a high radiation ~ Isolation Valves to the Containment monitor alarm on the air ejector exhaust, SI actuation would shut j these valves and depressing SI Reset would reopen them l i m
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- - ; : 2 of 2 aview of circe.ry for ventilation das;ses, cotors, and valves reported by EPCO resulted in discovery of similar designs in ESF-actuated components at i
culd return to its normal mode following the reset of an ESF s l irstcctive actions of the affected systems could be compromised once the ~, isscciated actuation signal is reset. These two plants had Stone and Webster ingineering Corporation for the architect-engineer as did the North Anna inits. [h2 Stone and Webster Engineering Corporation and VEPCO are preparing d i thangas to preclude safety-related equipment from moving out of its emargency node upon reset of an Engineered Safety Features Actuation Signal (ESFAS). This corrective action has been found acceptable by the NRC, in that, upoy res0t of ESFAS, all affected equipment remains in its emergency mode. M* y I reviews of selected areas of ESFAS reset action on PWR -f/.} 'acilities and, in some cases, this review was limited to examination of logic. 7heNRChasperformed e) It has been determined that logic diagrams may not i f .iagrams and procedures..dequately reflect as-built conditions; therefore, the reques l rawings must be done at the schematic /eler.antary diagram level. l ESF reset trhere have been several communicatio'ns to licensees from the NR \\ Beneric letters issued in November,1978 and October,1979 on containment Octions. i senting and purging during normal operation. Inspection and Enforcement sulletins Nos. 79-05, OSA, 058, 06A, 058 and 08 that addressed the events at i TMI-2 and NUREG-0578, TMI-2 Lessons Learned Task For_ce __ Status Report and i lications has Short-Term Recommendations. However, each of these
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We ar addressed only a limited area of the ESF's. reviews undertaken for this Bulletin address all of we ESF's. 1 l f l Actions To Be Taken By Licensees: lfor all PWR and BWR facilities with operating licenses: Review the drawings for all sys,tems serving safety-related functions at 1 the schematic level to determine whether or not upon the reset of an ESF '1. actuation signal, all associated safety-related equipment remains in its cmergency mode. Verify the actual installed instrumentation and controls a > 2. a test to demonstrate that all equipment remains in its emergency mode upon removal of the actuating signal and/or manual resetting of theProvide a M.- vcrious isolating or actuation signals. performance of the testing in your response to this Bulletin. c If any safety-related equipment does not remain in its emergency mode upo reset of an ESF signal at your facility, describe proposed system 3. d to rodification, design change, or other corrective action planne l l resolve the problem. i w-ww .-,n 4 ,---,-,--,e-w a ~, ,-m-m -m
arch 13, Ir.0 . in f:3. ?;-05 t e 3 cf 3 i d incin :: pert in writing within 90 days, the results of your rev ew ann item 3 i a list of all devices which respond as discussed d a schedule for '4. taken or planned to assure adequate equipment control, anThis information i.plementation of corrective action.Accordingly, you are requested to tt nts of the provisions of 10 CFr. 50.54(f). provide within the time period specified above, written s a eme Raports shall the above information, signed under oath or affircation. l Office and be submitted to the Of rector of the appropriate N i d Enforcement, i 20555. Div tsion of Reactor Operations Inspection, Washington, D.C. i Bulletin is Fcr all po er reactor facilities with a construction parmit, th s far information only and no written response is required. Approval was. 8130225 (R0072); clearance expires 7-31-80.ified generic problems. entq given under a blanket clearance specifically for ident Approved by GAO, t: ; 4 -8 9 O e m 1 a l -.a t ( ~ L i i I i l l
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l i 1 ENCLOSURE 4 Uh" ED STATES NUCLEAR kEGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 September 14, 1979 .- ;7 IE Information Notice No. 79-22 ~ QUALIFICATION OF CONTROL SYSTEMS ' ".Da ~' Pu511'c Service Electric and Gas Company notified the NRC of a potential unreviewed safety question at their Salem Unit i facility. This notification was based on a l I continuing revie. by Westinghouse of the environmental qualifications of equipment that they supply for nuclear steam supply systems. Based on the present status of this effert, Westinghouse has informed their customers that the performance 1 of non-safety grade equipment subjected to an adverse environoent could impact l the protective functions performed by safety grade equipment. These non-safetf Grade systems include: l Steam generator power operated relief valve control system Pressurizer power operated relief valve control systcr. Main feersater control system Automatic rod control system l Th se systems could potentially malfunction due to a high energy line break inside or outsidt of containment. NRC is also concerned that the adverse l environment could also give erroneous information to the plant operators. l L'astinghouse states that the consequences of such an event could possibly be l core limiting than results presented in Safety Analysis Reports, however, Westinghouse also states that the severity of the results can be limitad by operator actions together with operating characterisitics of the safety systems. Fur.ther, Westinghouse nas recommended to their customers that they I review their systems to cetermine whether any unreviewed safety questions exist. l This Information Notice is provided as an early notification of a possibly significant matter. It is expected that recipients will review the information for possibie applicability to their facilities. No specific action or response is requested at this time. If NRC evaluations so indicate, further licensee a:tions may be requested or required. If you have questiens regarding this matter, please contact the Director of the appropriate NRC Regional Office. 1 No written response to th'is Information Notice is required, l
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REPRINT Westinghouse Electric Corporation-Water Reactor Division Nuclear Service Division Box 2728 Pittsburgh, Pennsylvania 15230 August 30, 1979 -l PSE-79-21 Mr. F. P. Librizzi, General Manager Electric Production Public Service Electric and Gas Company 7;pj 80 Park Place i Newark, New Jersey 07101 ' 71 j
Dear Mr. Librizzi:
Public Service Electric and Gas Co. 2 Salem Unit Nn.1 QUALIFICATION OF CONTkOL SYSTENS As part of a continuing review of the environmental qualifications of Westinghouse supplied NSSS equipment, Westinghouse has also found it necessary to consider the interaction with non-safety grade systems. This investigation has been conducted to determine if the performance of non-safety grade systems which may not be protected from an adverse environment could impact the protective functions performed by NSSS e safety grade equipment. The NSSS control and protection systems were j included in this review to assess the adequacy of the present environ- } mental qualification requirements, y As a result of this review, several systems were identified which, if subjected to an adverse environment, could potentially lead to control system operation which may impact protective functions. These systems Lre: Steam generator pcwer operated relief valve control system Pressurizer power operated relief valve control system Main feedwater control system Automatic rod control system ~ ,,,-..,,-,.,-----,-,,----e-
l Page 2 PSE-79-21 s, Each of the above mentioned systems could potentially malfunction if impacted by adverse environments due to a high energy line break inside ~ In each case, a limited set of breaks, coupled or outside containment. with possible consequential control malfunction in an adverse direction, of the above events sould yield results which are more limiting than those In all cases, however, the presented in the plant Safety Analysis Reports. severity of the results can be limited by operator actions together with operating characteristics of the safety systems. _, i We believe these systems identified do not constitute a substantial safety ' ~j Y'l However, Westinghouse recommends you review them to determine if hazard. any unreviewed safety questions or significant deficiencies exist in your 1 pl ant ( s). To assist you in understanding these concerns Westinghouse will hold a seminar in Pittsburgh on Thursday, September 6 at Westinghouse R&O Center, The seminar will Building 701, with all our operating plant customers. address the potential impact of these concerns for various plant designs and various licensing bases. Please contact your WNSD Regional Service office to confirm your attendance We will provide additional details concerning the agenda at the seminar. and other meeting arrangements as they become available. Very truly yours. l ORIGINAL SIGNED BY j F. Noon, Manager Eastern Regional & WNI Support SR4/CCl3&l4 cc: H. J. Midura H. J. Heller R. D. Rippe T. N. Taylor R. A. Uderitz C. F. Barclay W n,, ,,-,. -.., -,.. - ~., - -., n. v-.,. ..e,_-,-.
i ~ _. -. _. _ _ _ _ _ _. _ _ _ _ _, REPRINT c PUBLIC SERVICE ELECTRIC AND GAS COMPANY Salem Nuclear Generating Station ~ P. O. Box 56 Hancocks Bridge, New Jersey 08033 'a September 10, 1979 Mr. Boyce H. Grier* Director of USNRC Office of Inspection and Enforcement Region I .I 631 Park Avenue King of Prussia, Pennsylvania 19406 ,3 '
Dear Sir:
s REPORTABLE OCCURRENCE 79-58/OlP ~ SALEM NO. 1 UNIT LER This letter will serve to confirm our telephone repo% to Mr. Gary Schneider of the Regional NRC office on Friday, September 6,1979, advising of a potential reportable occurrence in accordance with Technical Specification 6.9.1.8. We have been notified by our Engineering Department that a Westing-house conducted review of the environmental qualifications of Westinghouse supplied NSSS equipment has identified that conditions associated with high energy line breaks insice or outside containment j and their impact on non-safety control systems may constitute an c unreviewed safety question. The control systems concerned are steam j l gener ator power operated relief ve,;ve control, pressurizer power ] l operated relief valve control, main feedwater control and automatic rod control systems. A detailed report will be submitted in the time period specified by the Technical Specifications. Very truly yours, Original Signed By H. J. Midura Manager - Salem Generating Station AWK:jds CC: General Manager - Electric Production i l Manager - Quality Assurance l l Q U P E ' W Y 7 1 T 1' $ sl $ S -}}